ML20138D765
| ML20138D765 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 11/21/1985 |
| From: | Elin J, Morrill P, Pate R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20138D684 | List: |
| References | |
| 50-275-OL-85-03, 50-275-OL-85-3, NUDOCS 8512130294 | |
| Download: ML20138D765 (53) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION V Report No. 50-275/0L-85-03 Docket Nos. 50-275 and 50-323 Licensee: Pacific Gas and Electric Company 77.Beale Street San Francisco, California 94106 Facility Name: Diablo Canyon Units.1 and 2 Examinations at: Avila Beach, California-Examination conducted: November 13, 198 Examiners:
b P. J.181orrill Date Signed
- Y Y g . O. Elin " Dhte S'igned Approved ~by: .
< N /7 g J. Pate, Chief, Operations Section Date Sighed Summary:
Examinations were conducted on November 13, 1985. The written examination was administered on November 13, 1985 to one senior reactor operator retake candidate (SRO) who passed. In addition, the oral and simulator examinations were administered to two SRO candidates. One candidate's exam was a continuation of the examination given in May 1985. Both of these candidates passed the oral and simulator examinations.
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REPORT DETAILS
- 1. Examiners:
P. Morrill, Chief Examiner, Region V J. Elin, Region V-C. Shiraki, HQ, Auditor
- 2. Persons Attending the Exit Meeting:
l November 14, 1985 NRC i P. Morrill, Region V J. Elin, Region V l C. Shiraki, NRC HQ Pacific Gas and Electric Company R. Patterson, Plant Superintendent T. Martin, Training Manager R. Fisher, Senior Operations Engineer R. Graham, Senior Training Instructor W. Kaefer, Assistant Plant Manager / Support Services ;
J. Molden, Operations Training Supervisor ,
- 3. Facility Review of the Written Examination At the conclusion of the written examinations,.the, examiners met with R. Graham, and J. Molden of'the-Training Department to review the written examinations and answer keys. Their comments were incorporated l in to the master examination keys prior to grading the candidates' responses. .
- 4. Exit Meetings 't a= ,
Operating Examinations ,
Simulator and operating examinations were' '
conducted ~on November l13, 1985.
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No weaknesses were observed. . '+< ,
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Clear Passes ~
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The names of the candidates whoLelearly pass $dlboth the simulator and
. operating examinations were provided 'at the exit meeting. ' 't .
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U. 5. NUCLEAR REGULATORY COMMISSION --
SENIOR REACTOR OPERATOR EXAMlHATION raci11ty: D/MLO 09 WON / fA Reactor Type: lk/EST/A)(~/}DU.se PotIR Date Administered: /YOl/6/F)/[d /,8, /YcP,5 Examiner: k.M M/L -
Candidate:
INSTRUCTIONS TO CANDIDATE: ,
Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheet. Points for each question are indicated in parenthesis after the question. The passing grade requires at least 70% in ea Examination papers will be pick,ch category ed up six (6) and hoursa final aftergrade of at least 80%.
tiie examination starts.
Category % of Candidate's % of Value Total Score Cat. Value Catecory
- 5. Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics
- 6. Plant Systems Design, Edntrol and Instrumentation
- 7. Procedures - Normal, Abnormal.
Emergency and Radiolocical Control
- 8. Administrative Procedures, Conditions, and Limitations TOTALS Final Grade fl work done on this examination is my own; I have neither given nor received alc. .
O Canc1date's Signature mAsreR key _
e m y4 ,,.
. 4 EQUATION SHEET f = ma v = s/t w = mg 2 Cycle efficiency = Net Work (out) a = v,t + at Energy (in)
E = mC a = (v - y )/t
> f o -At KE = mv v g =v + at A = AN A = A,e j PE = mgh to = 0/t A = in 2/tg = 0.693/tg W = vaP AE = 931Am tgeff)=(t)(t)
- u. u (c +g)
Q = $CPAT I Ie -EX
, o Q = UAAT I.Ie S Pwr = W g$
~
I=I o 10
- a P=P 10 SUR(t) TVL = 1.3/p P=P et/T HVL = 0.693/p o
SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg)
SUR = 26
!A*ffD) CR x = S/(1 - K,gg )
g,p T = '(1*/p ) + [(f.'p)/Aeff] p
~
b" b T = 1*/ (p - T) M = 1/(1'- K,gg) = CR /CR0 g
"( ~ 0)! eff P M = (1 - K,gg)0 (I ~ eff)1 P " ( eff~ )! eff " A eff/Keff SDM = (1 - K,gg)/K,gg p= [L*/TK,'gg] + [B/(1 + A,ggT )] t* = 1 x 10~ seconds y -1 P = I(V/(3 x 1010) eff
= 0.1 seconds I = No Id 1y =Id 22 i
WATER PARAMETERS Id g
=Id 2 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters)
I gal. = 3.78 liters R/hr = 6 CE/d (feet)
I ft = 7.48 gal. MISCELLANEOUS CONVERSIONS ,
3 10 Density = 62.4 lbm/ft , 1 Curie = 3.7 x 10 dpa Density = 1 gm/cm 1 kg = 2.21 lbm 3
Heat of vaporization = 970 Btu /lbm I hp = 2.54 x 10 BTU /br 0
i Feat of fusion = 144 Btu /lbm 1 Mw = 3.41 x 10 Stu/hr 1 Atm = 14.7 psi = 29.9 in. Hg. 1 Btu = 778 ft-lbf 2
I ft. H O = 0.4335 lbf/in 1 inch - 2.54 cm 2
F = 9/5 C + 32
~
C = 5/9 ( F - 32)
~
t Properties of Saturated Steam and Saturated Water
- Absolute Pressure Vacuum Temper- Heat of Latent IIcat Total Heat Specific Volume inches Inches ature the of , of Steam y Lbs.
Sq. N.r of Hg of Hg Liqu d Evaporation t be Water Steam P' o,. . r. s oris. seufis. s.o ris. cu.re.n,,sw cu.re. n.,is.
0.0087 0.02 29.90 32.018 0.0003 1075.5 1075.5 0.016022 3302.4 0.10 0.20 29.72 35.023 3.026 1073.8 1076.8 0.016020 2945.5 0.15 0.31 29.61 45.453 13.498 1067.9 1081.4 0.016020 2004.7 0.20 0.41 29.51 53.160 21.217 1053.5 1084.7 0.016025 1526.3 0.25 0.51 29.41 59.323 27.382 1000.1 1087.4 0.016032 1235.5 0.30 0.61 29.31 64.484 32.541 1057.1 1089.7 0.016040 1039.7 0.35 0.71 29.21 68.939 36.992 1054.6 1091.6 0.016048 898.6 0.40 0.81 29.11 72.869 40.917 1052.4 1093.3 0.016056 792.1 0.45 0.92 29.00 76.387 44.430 1050.5 1094.9 0.016063 708.8 0.50 1.02 28.90 79.586 47.623 1048.6 10 % .3 0.016071 641.5 0.60 1.22 28.70 85.218 53.245 1045.5 1098.7 0.016085 540.1 0.70 1.43 28.49 90.09 58.10 1042.7 1100.8 0.016099 466.94 0.80 1.63 28.29 94.38 62.39 1040.3 1102.6 0.016112 411.69 0.90 1.83 28.09 98.24 66.24 1038.1 1104.3 0.016124 368.43 1.0 2.04 27.88 101.74 69.73 1036.1 1105.8 0.016136 333.60 1.2 2.44 27.48 107.91 75.90 1032.6 1108.5 0.016158 280.%
1.4 2.85 27.07 113.26 81.23 1029.5 1110.7 0.016178 243.02
, 1.6 3.26 26.66 117.98 85.95 1026.8 1112.7 0.0161 % 214.33 1.8 3.66 26.26 122.22 90.18 1024.3 1114.5 0.016213 191.85 2.0 4.07 25.85 126.07 94.03 1022.1 1116.2 0.016230 173.76 2.2 4.48 25.44 129.61 97.57 1020.1 1117.6 0.016245 158.87 2.4 4.89 25.03 132.88 100.84 1018.2 1119.0 , 0.016260 146.40 2.6 5.29 24.63 135.93 103.88 1016.4 1120.3 0.016274 135.80 2.8 7.70 24.22 138.78 106.73 1014.7 1121.5 0.016287 126.67 3.0 6.11 23.81 141.47 109.42 1013.2 1122.6 0.016300 118.73 3.5 7.13 22.79 147.56 115.51 1009.6 1125.1 0.016331 102.74 4.0 8.14 21.78 152.% 120.92 1006.4 1117.3 0.016358 90.64 4.5 9.16 20.76 157.82 125.77 1005.5 1129.3 0.016384 83.03 5.0 10.18 19.74 162.24 130.20 1000.9 1131.1 0.016407 73.532 5.5 11.20 18.72 166.29 134.26 998.5 1132.7 0.016430 67.249 6.0 12.22 17.70 170.05 138.03 996.2 1134.2 0.016451 61.984 6.5 13.23 16.69 173.56 141.54 994.1 1135.6 0.016472 57.5 %
7.0 14.25 15.67 176.84 144.83 992.1 1136.9 0.016441 53.650 7.5 15.27 14.65 179.93 147.93 990.2 1138.2 0.016510 50.294 8.0 16.29 13.63 182.86 150.87 988.5 1139.3 0.016527 47.345 8.5 17.31 12.61 185.63 153165 986.8 1140.4 0.016545 44.733 9.0 18.31 11.60 188.27 156.30 985.1 1141.4 0.016561 42.402 9.5 19.34 10.58 190.80 158.84 983.6 1142.4 0.016577 40.310 10.0 20.36 9.56 193.21 161.26 982.1 1143.3 0.016592 38.420 11.0 22.40 7.52 197.75 165.82 979.3 1145.1 0.016622 35.142 12.0 24.43 5.49 201.% 170.05 976.6 1146.7 0.016650 32.394 13.0 26.47 3.45 205.88 174.00 974.2 1148.2 0.016676 30.057 14.0 38.50 1.42. 209.56 177.71 971.9 1149.6 0.016702 28.043 Pressure Temper- Heat of Latent Heat Total Heat Specific Volume Lbs. per Sq. In. the of Steam Absolute Gage ature Liquid of Evaporation y P'
t he Water Steam P o e,- r. scuris. seufis. scu ris. cu. te. ,., is. co. rt. ,,, is.
14.6 % 0.0 212.00 180.17 970.3 1150.5 0.016719 26.799 15.0 0.3 213.03 181.21 %9.7 1150.9 0.016726 26.290 16.0 1.3 216.32 184.52 %7.6 1152.1 0.016749 24.750 17.0 2.3 219.44 187.66 %5.6 1153.2 0.016771 23.385 18.0 3.3 222.41 190.66 %3.7 - 1154.3 0.016793 22.168 19.0 4.3 225.24 193.52 %I .8 1155.3 0.016814 21.074 20.0 5.3 227.% 1%.27 960.1 1156.3 0.016834 20.087 21.0 6.3 230.57 198.90 958.4 1157.3 0.016854 19.190 22.0 7.3 233.07 201.44 956.7 1158.1 0.016873 18.373 23.0 8.3 235.49 203.88 955.1 1159.0 0.016891 17.624 24.0 9.3 237.82 206.24 953.6 1159.8 0.016909 16.936 25.0 10.3 240.07 208.52 952.1 1160.6 0.016927 16.301 26.0 11.3 242.25 210.7 950.6 1161.4 0.016944 15.7138 27.0 12.3 244.36 212.9 949.2 1162.1 0.016 % 1 15.1684 28.0 13.3 246.41 214.9 947.9 1162.8 0.016977 14.6607 29.0 14.3 248.40 217.0 946.5 1163.5 0.016993 14.1869 30.0 15.3 250.34 218.9 945.2 !!64.1 0.017009 13.7436 31.0 16.3 252.22 220.8 943.9 1164.8 0.017024 13.3280 32.0 17.3 254.05 222.7 942.7 1165.4 0.017039 12.9376 33.0 18.3 255.84 224.5 941.5 1166.0 0.017054 12.5700 34.0 19.3 257.58 226.3 940.3 1166.6 0.017069 12.2234 l
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PNperties of Saturated Steam and Saturated Water-cc,ntinued Pressure Temper- Heat of Latent Heat Total Heat Specific Volume Lbs. per Sq. In. sture the of of Steam Gage Liquid Evaporation p Absol,ute !
P P Water i Steam rwar . F. aiu tid. aiutis. siu ris. cu.ri. p.,is. I cu.re.p.,is.
35.0 20.3 259.29 228.0 939.1 1167.1 0.017083 11.8959 36.0 21.3 260.95 229.7 938.0 1167.7 0.017097 11.5860 37.0 22.3 262.58 231.4 936.9 1168.2 0.017111 11.2923 38.0 23.3 264.17 233.0 935.8 1168.8 0.017124 11.0136 39.0 24.3 265.72 234.6 934.7 1169.3 0.017138 10.7487 40.0 25.3 267.25 230.1 933.t> 1169.8 0.017151 10.4965 41.0 26.3 268.74 237.7 932.6 1170.2 0.017164 10.2563 42.0 27.3 270.21 239.2 931.5 1170.7 0.017177 10.0272 43.0 28.3 271.65 240.6 930.5 1171.2 0.017189 9.8083 44.0 29.3 273.06 242.1 929.5 1171.6 0.017202 9.5991 45.0 30.3 274.44 243.5 928.6 1172.0 0.017214 9.3988 46.0 31.3 275.80 244.9 927.6 1172.5 0.017226 9.2070 47.0 32.3 277.14 246.2 926.6 1172.9 0 017238 9.0231 48.0 33.3 278.45 247.6 925.7 1173.3 0.017250 8.8465 49.0 34.3 279.74 248.9 924.8 1173.7 0.017262 8.6770 50.0 35.3 281.02 250.2 923.9 1174.1 0.017274 8.5140 51.0 36.3 282.27 251.5 923.0 1174.5 0.017285 8.3571 52.0 37.3 283.50 252.8 922.I 1174.9 0.0172 % 8.2061 53.0 38.3 284.71 254.0 921.2 1175.2 0.017307 8.0606 54.0 39.3 285.90 255.2 920.4 1175.6 0.017319 7.9203 55.0 40.3 287.08 256.4 919.5 1175.9 0.017329 7.7850 56.0 41.3 288.24 257.6 918.7 1176.3 0.017340 7.6543 57.0 42.3 289.38 258.8 917.8 1176:6 0.017351 7.5280 58.0 43.3 290.50 259.9 9I7.0 1I77.0
- 0.017362 7.4059 59.0 44.3 291.62 261.1 916.2 1177.3 Q.017372 7.2879 60.0 45.3 292.71 262.2 915.4 1177.6 0.01/383 61.0 7.1736 46.3 293.79 263.3 914.6 1177.9 0.017393 7.0630 62.0 47.3 294.86 264.4 913.8 1178.2 0.017403 6.9558 63.0 48.3 295.91 265.5 913.0 1178.6 64.0 0.017413 6.8519 49.3 2%.95 266.6 912.3 1178.9 0.017423 6.7511 65.0 50.3 297.98 267.6 911.5 1179.1 66.0 0.017433 6.6533 51.3 298.99 268.7 910.8 1179.4 0.017443 6.5584 67.0 52.3 299.99 269.7 910.0 1179.7 68.0 0.017453 6.4662 53.3 300.99 270.7 909.3 1180.0 0.017463 69.0 54.3 6.3767 301.% 271.7 908.5 1180.3 0.017472 6.28 %
70.0 55.3 302.93 272.7 907.8 1180.6 71.0 0.017482 6.2050 56.3 303.89 273.7 907.1 1180.8 0.017491 72.0 57.3 304.83 6.1226 274.7 906.4 1181.1 0.017501 6.0425 73.0 58.3 305.77 74.0 59.3 306.69 275.7 . 405.7 1181.4 0.017510 5.9645 276.6 905.0 1181.6 0.017519 5.8885 75.0 60.3 307.61 277.6 904.3 76.0
!!81.9 0.017529 5.8144 61.3 308.51 278.5 903.6 1882.1 0.017538 77.0 62.3 5.7423 309.41 279.4 902.9 1182.4 0.G17547 78.0 63.3 5.6720 310.29 280.3 902.3 1182.6 0.017556 79.0 64.3 311.17 5.6034 281.3 901.6 1182.8 0.017565 5.5364 80.0 65.3 312.04 282.1 900.9 1183.1 0.017573 5.4711 81.0 66.3 312.90 283.0 900.3 82.0 1183.3 0.017582 5.4074 67.3 313.75 283.9 899.6 1183.5 0.017591 83.0 68.3 314.60 5.3451 284.8 899.0 1183.8 0.017600 5.2843 84.0 69.3 315.43 285.7 898.3 1184.0 0.017608 5.2249 85.0 70.3 316.26 286.5 897.7 86.0 1184.2 0.017617 5.1669 71.3 317.08 287.4 897.0 1184.4 0.017625 87.0 72.3 317.89 5.110!
88.0 288.2 8%.4 1184.6 0.017634 5.0546 73.3 318.69 289.0 895.8 1184.8 89.0 74.3 0.017642 5.0004 319.49 289.9 895.2 1185.0 0.017651 4.9473 90.0 75.3 320.28 290.7 894.6 91.0 1885.3 0.017659 4.8953 76.3 321.06 291.5 893.9 1185.5 92.0 77.3 0.017667 4.8445 321.84 292.3 893.3 1185.7 0.017675 4.7947 93.0 78.3 322.61 293.1 892.7 94.0 1185.9 0.017684 4.7459 79.3 323.37 293.9 892.I i186.0 0.017692 4.6982 95.0 80.3 324.13 294.7
%.0 891.5 1186.2 0.017700 4.6514 81.3 324.88 295.5 891.0 1186.4 97.0 82.3 0.017708 4.6055
'325.63 2%.3 890.4 1I86.6 0.017716 98.0 83.3 326.36 4.56 %
297.0 '889.8 1186.8 0.017724 4.5166 99.0 84.3 327.10 297.8 889.2 1187.0 0.017732 4.4734 100.0 85.3 327.82 298.5 888.6 1187.2 0.017740 4.4310 101.0 86.3 328.54 299.3 888.1 1187.3 0.01775 4.3895 102.0 87.3 329.26 300.0 887.5 103.0 1187.5 'O.01776 4.3487 88.3 329.47 300.8 886.9 1187.7 104.0 89.3 0.01776 4.3087 .
330.67 30I.5 886.4 1187.9 0.01777 105.0 4.2695 90.3 331.37 302.2 885.8 1888.0 106.0 91.3 0.01778 - 4.2309 332.06 303.0 885.2 1188.2 0.01779 107.0 92.3 332.75 4.1931 303.7 884.7 1188.4 0.01779 4.1560 108.0 93.3 333.44 304.4 884.1 109.0 1188.5 0.01780 4.1195 94.3 . 334.11 305.1 883.6 1188.7 0.01781 4.0837 n_
~
Properties of Saturated Steam and Saturated Water-continued Pressure Temper- IIcat of L Specific Volume Lbs. per Sq. In. ature the ! atent of IIcat l Totalllent of Steam p Absolute Gage r Liquid ! Evaporation l W,at er Steam P' P o,er . F. Biorih. I niuris. niuris. cu ri. n , is. c u. ei. ner is.
110.0 95.3 334.79 305.8 883.1 1188.9 0.01782 4.0484 111.0 %.3 335.46 306.5 882.5 1189.0 0.01782 4.0138 112.0 97.3 336.12 307.2 882.0 1889.2 0.01783 3.9798 113.0 98.3 336.78 307.9 881.4 1I89.3 0.01784 3.9464 114.0 99.3 337.43 308.6 880.9 1189.5 0.01785 3.9136
!!5.0 100.3 338.08 309.3 880.4 1889.6 0.01785 3.8813 116.0 101.3 338.73 309.9 879.9 1189.8 0.01766 3.8495 117.0 102.3 339.37 310.6 879.3 1189.9 0.01787 3.8183 118.0 103.3 340.01 311.3 878.8 1190.1 0.01787 3.7875 119.0 104.3 340.64 311.9 878.3 1190.2 0.01788 3.7573 120.0 105.3 341.27 312.6 877.8 1190.4 0.01789 3.7275 121.0 106.3 341.89 313.2 877.3 1190.5 0.01790 3.6983 122.0 107.3 342.51 313.9 876.8 1190.7 0.01790 3.6695 123.0 108.3 343.13 314.5 876.3 1190.8 0.01791 3.6411 124.0 109.3 343.74 315.2 875.8 1190.9 0.01792 3.6132 125.0 110.3 344.35 315.8 875.3 1191.1 0.01792 3.5857 126.0 111.3 344.95 316.4 874.8 1191.2 0.01793 3.5586 127.0 112.3 345.55 317.1 874.3 1191.3 0.01794 3.5320 128.0 113.3 346.15 317.7 873.8 1191.5 0.01794 3.5057 129.0 114.3 346.74 318.3 873.3 1191.6 0.01795 3.4799
- 130.0 115.3 347.33 319.0 872.8 1191.7 0.017 % 3.4544 131.0 116.3 347.92 319.6 872.3 1191.9 0.01797 3.4293 132.0 117.3 348.50 320.2 871.8 -1192.0 0.01797 3.4046 133.0 118.3 349.08 320.8 871.3 1192.1 0.01798 3.3802 134.0 119.3 349.65 321.4 870.8 1192.2 0.01799 3.3562 135.0 120.3 350.23 322.0 870.4 1192.4 0.01799 3.3325 136.0 121.3 350.79 322.6 869.9 1192.5 0.01800 3.3091 137.0 122.3 351.36 323.2 869.4 1192.6 0.01801 3.2861 138.0 123.3 351.92 323.8 868.9 - 1192.7 0.01801 3.2634 139.0 124.3 352.48 324.4 868.5 1192.8 0.01802 3.2411 140.0 125.3 353.04 325.0 868.0 1193.0 0.01803 3.2190 141.0 126.3 353.59 325.5 867.5 1193.1 0.01803 3.1972 142.0 127.3 354.14 326.1 867.1 1193.2 0.01804 3.1757 143.0 128.3 354.69 326.7 866.6 1193.3 0.01805 3.1546 144.0 129.3 355.23 327.3 866.2 1193.4 0.01805 3.1337 145.0 130.3 . 355.77 327.8 865.7 4193.5 0.01806 3.1130 146.0 131.3 356.31 328.4 865.2 1193.6 0.01806 3.0927 147.0 132.3 356.84 329.0 864.8 1193.8 0.01807 3.0726 148.0 133.3 357.38 329.5 864.3 1193.9 0.01808 3.0528 149.0 134.3 357.91 330.I' 863.9 1194.0 0.01808 3.0332 150.0 135.3 358.43 330.6 863.4 1194.1 0.01809 3.0139 152.0 137.3 359.48 331.8 862.5 1194.3 0.01810 2.9760 154.0 139.3 360.51 332.8 861.6 1194.5 0.01812 2.9391 156.0 141.3 361.53 333.9 860.8 1194.7 0.01813 2.9031 158.0 143.3 362.55 335.0 859.9 1194.9 0.01814 2.8679 160.0 145.3 363.55 336.1 859.0 1195.1 0.01815 2.8336 162.0 147.3 364.54 337.1 858.2 1195.3 0.01817 2.8001 164.0 149.3 365.53 338.2 857.3 '1195.5 0.01818 2.7674 166.0 151.3 366.50 339.2 856.5 1195.7 0.01819 2.7355 168.0 153.3 367.47 340.2 855.6 1195.8 0.01820 2.7043 170.0 155.3 368.42 341.2 854.8 11 % .0 0.01821 2.6738 172.0 157.3 369.37 342.2 853.9 11 % .2 0.01823 2.6440 174.0 159.3 370.31 343.2 853.1 11 % .4 0.01824 2.6149 176.0 161.3 371.24 344.2 852.3 11 % .5 0.01825 2.5864 178.0 163.3 372.16 345.2 851.5 11 % .7 0.01826 2.5585 180.0 165.3 373.08 346.2 850.7 11 % .9 0.01827 2.5312 182.0 167.3 373.98 347.2 849.9 1197.0 0.01828 2.5045 184.0 169.3 374.88 348.1 849.1 1197.2 0.01830 2.4783 186.0 171.3 375.77 349.1 848.3 1197.3 0.01831 2.4527 188.0 173.3 376.65 350.0 847.5 1197.5 0.01832 2.4276 190.0 175.3 377.53 350.9 846.7 1197.6 0.01833 2.4030 192.0 177.3 378.40 351.9 845.9 1197.8 0.01834 2.3790 194.0 179.3 379.26 352.8 845.1 1197.9 0.01835 2.3554 1%.0 181.3 380.12 353.7 -
844.4 1198.1 0.01836 2.3322 198.0 183.3 380.% 354.6 843.6 1198.2 0.01838 2.3095 200.0 185.3 381.80 355.5 842.8 1198.3 0.01839 2.28728 205.0 190.3 383.88 357.7 840.9 1198.7 0.01841 2.23349 210.0 195.3 385.91 359.9 839.1 1199.0 0.01844 2.18217 215.0 200.3 387.91 362.1 837.2 1199.3 0.01847 2.13315 220.0 205.3 389.88 364.2 835.4 1199.6 ' O.01850 2.08629 225.0 210.3 391.80 366.2 833.6 1I99.9 0.01852 2.04143 230.0 215.3 393.70 368.3 831.8 1200.1 0.01855 1.99846 235.0 220.3 395.56 370.3 830.1 1200.4 0.01857 1.95725 240.0 225.3 397.39 372.3 828.4 1200.6 0.01860 1.91769 245.0 230.3 399.19 374.2 826.6 1200.9 0.01863 1.87970
Properties of Saturated Steam and Saturated Water-concluded Pressure Temper- IIcat of Latentlicat Total IIcat Specific Volume Lbs. per Sq. In. the of ature Liquid Evaporation of Steam p Absolute Gage r h8 P, P Water Steam o .. r. acuris. niu ris. scutis. co. re. r., is. cu. re. c., is.
250.0 235.3 400.97 376.1 8iST0 1201.1 0.01865 1.84317 255.0 240.3 402.72 378.0 823.3 1201.3 0.01868 1.80802 260.0 245.3 404.44 379.9 821.6 1201.5 0.01870 1.77418 265.0 250.3 406.13 381.7 820.0 1201.7 0.01873 1.74157 270.0 255.3 407.80 383.6 818.3 1201.9 0.01875 1.71013 275.0 260.3 409.45 385.4 816.7 1202.1 0.01878 1.67978 280.0 265.3 411.07 387.1 815.1 1202.3 0.01880 1.65049 285.0 270.3 412.67 388.9 813.6 1202.4 0.01882 1.62218 290.0 275.3 414.25 390.6 812.0 1202.6 0.01885 1.59482 295.0 280.3 415.81 392.3 810.4 1202.7 0.01887 1.56835 300.0 285.3 417.35 394.0 808.9 1202.9 0.01889 1.54274 320.0 305.3 423.31 400.5 802.9 1203.4 0.01899 1.44801 340.0 325.3 428.99 406.8 797.0 1203.8 0.01908 1.36405 360.0 345.3 434.41 412.8 791.3 1204.1 0.01917 1.28910 380.0 365.3 439.61 418.6 785.8 1204.4 0.01925 1.22177 400.0 385.3 444.60 424.2 780.4 1204.6 0.01934 1.16095 420.0 405.3 449.40 429.6 775.2 1204.7 0.01942 1.10573 440.0 425.3 454.03 434.8 770.0 1204.8 0.01950 1.05535 460.0 445.3 458.50 439.8 765.0 1204.8 0.01959 1.00921 480.0 465.3 462.82 444.7 760.0 1204.8 0.01 % 7 0.96677 500.0 485.3 467.01 449.5 755.1 1204.7 0.01975 0.92762 520.0 505.3 471.07 454.2 750.4 1204.5 0.01982 0.89137 540.0 525.3 475.01 458.7 745.7 1204.4 0.01990 0.85771 560.0 545.3 478.84 463.1 741.0 1204.2 0.01998 0.82637 580.0 565.3 482.57 467.5 736.5 1203.9 'O.02006 0.79712 600.0 585.3 486.20 471.7 732.0 1203.7 0.02013 0.76975 623.0 605.3 489.74 475.8 727.5 1203.4 0.02021 0.74408 640.0 625.3 493.19 479.9 723.1 1203.0 0.02028 0.71995 660.0 645.3 4 %.57 483.9 718.8 1202.7
- 0.02036 0.69724 680.0 665.3 499.86 487.8 714.5 1202.3 0.02043 0.67581 700.0 685.3 503.08 491.6 710.2 1201.8 0.02050 0.65556 720.0 705.3 5 % .23 495.4 706.0 1201.4 0.02058 0.63639 740.0 725.3 509.32 499.1 701.9 1200.9 0.02065 0.61822 760.0 745.3 512.34 502.7 697.7 1200.4 0.02072 0.60097 780.0 765.3 515.30 506.3 693.6 1199.9 0.02080 0.58457 800.0 785.3 518.21 509.8 689.6 !!99.4 0.02087 0.568 %
820.0 805.3 521.06 513.3 685.5 1198.8 0.02094 0.55408 840.0 825.3 523.86 516.7 681.5 1198.2 0.02101 0.53988 800.0 845.3 526.60 520.1 677.6 1197.7 0.02109 0.52631 880.0 865.3 529.30 523.4 673.6 1197.0 0.02116 0.51333 900.0 885.3 531.95 526.7 669.7 !!%.4 0.02123 0.50091 920.0 905.3 534.56 530.0 665.8 1195.7 0.02130 0.48901 940.0 925.3 537.13 533.2 661.9 1195.1 0.02137 0.47759 960.0 945.3 539.65 536.3 658.0 1194.4 0.02145 0.46662 980.0 %5.3 542.14 539 5 654.2 ; 1193.7 0.02152 0.45609 1000.0 985.3 544.58 542.6 650.4 1192.9 0.02159 0.445 %
1050.0 1035.3 550.53 550.1 640.9 1191.0 0.02177 0.42224 1100.0 1085.3 556.28 557.5 631.5 1189.1 0 02195 0.40058 1150.0 1135.3 561.82 564.8 622.2 1187.0 0.02214 0.38073 1200.0 1185.3 567.19 571.9 613.0 1184.8 0.02232 0.36245 1250.0 1235.3 572.38 578.8 603.8 1182.6 0.02250 0.34556 1300.0 1285.3 577.42 585.6 594.6 1180.2 0.02269 0.32991 1350.0 1335.3 582.32 592.2 585.6 1177.8 0.02288 0.31536 1400.0 1385.3 587.07 598.8 567.5 1175.3 0.02307 1450.0 0.30178 1435.3 591.70 605.3 567.6 1172.9 0.02327 0.28909 1500.0 1485.3 5 %.20 611.7 558.4 1170.1 0.02346 0.27719 1600.0 1585.3 604.87 624.2 540.3 1164.5 0.02387 0.25545 1700.0 1685.3 613.13 636.5 522.2 1158.6 0.02428 1800.0 0.23607 1785.3 621.02 648.5 503.8 1152.3 0.02472 0.21861 1900.0 1885.3 628.56 660.4 485.2 1145.6 0.02517 0.20278 2000.0 19MS.3 635.80 . 672.1 466.2 1138.3 0.02565 0.18831 2100.0 2085.3 642.76 683.8 446.7 1130.5 0.02615 0.17501 2200.0 2185.3 649.45 695.5 , 426.7 1122.2 0.02669 2300.0 0.16272 2285.3 655.89 707.2 406.0 1113.2 0.02727 0.15133 2400.0 23MS.3 662.11 719.0 384.8 1103.7 0.02790 0.14076 2500.0 2455.3 668.ll 731.7 361.6 1093.3 0.02859 0.13068 2600.0 2585.3 673.91 744.5 337.6 1082.0 0.02938 0.12110 2700.0 2685.3 679.53 757.3 312.3 1069.7 2800.0 0.03029 0.11194 2785.3 684.% 770.7 285.1 1055.8 0.03134 0.10305 1900.0 2885.3 690.22 785.1 254.7 1039.8 0.03262 0.09420 3000.0 2985.3 695.33 801.8 218.4 1020.3 3100.0 0.03428 - 0.0s500 3085.3 700.28 824.0 169.3 993.3 0.07452 3200.0 ~ 0.03681 31M5.3 705.08 875.5 56.1 931.6 0.04472 0.05663 3208.2 3193.5 705.47 906.0 0.0 906.0 0.05078 0.05078 TL
t sro5 5.01 (3.0)
You are supervising taking Unit 2 from hot standby to 20 % power l
using Operating Procedure OP L-2 (Hot Standby to Minimum Load).
For calculations show your assumptions and your work.
(a) What indication do you expect to see from the nuclear instrumentation which indicates that the reactor is critical ? (1.0)
(b) You observe that the source range instruments indicate 20 counts when all control banks are inserted (assume K(eff) = 0.935) and 400 counts when control bank D is at zero steps. (All other rods are fully withdrawn.) What is K(effective) at this point ? (1.0)
(c) The procedure requires that a start up rate of
+0.75 DPM be established. How much reactivity (PCM) is required ? (assume l$=0.007andA=0.1)
(a) The reactor is critical (or slightly supercritical) when source range instruments indicate a constant positive start-up rate and count rate cteadily increases with no reactivi ty being added to the core. (no rod motion)
(b) CR1(1 - K(off))1 = CR2(1 - K(eff))2 20(1 - 0.935) = 400(1 - K(eff))
1.3 = 400 - 400*k(eff) : 398. 7 = 400*K (ef f )
K(eff) = 398.7/400 : Kfgif1 = 01 22625 (c) SUR = 26.06/T since SUR = +0.75 T = 34.75 Sec.
and T = 0
([-/)/gf co 34.75 = (0.007-fD)/O.1/2 : 3.475 = 0.007 - [3 4.5 f = 0.007 = 700 PCM : therefore f 5 15) PCtj
References:
Operating Procedure DP O-2, Hot standby to Minimum Load Westinghouse, Reactor Core Control for Large Pressurized Water Reactorc, pages 9 9-17 1
1 O
)
i 5.02 (2.0)
Refer to Figure 5.1 which shows a very rapid positive tgattiyity I insertion into an already critical reactor (at 10-8 amps) at time t = 0. After a stable reactor period is reached, an equal and opposite negative tgactiyity insertion is made at t = 5 minutes (thus rendering the reactor just critical again). Assune that source neutrons are not significant in the transient, and that the reactivity from all sources is as shown.
(a) Show the resulting startup rate as a function of time for the reactivity changes shown. (1.0)
(b) Show the reactor power level as a function of time for the reactivity changes shown. (1.0) 1 (a) See attached graph (b) See attached graph l
Reference:
Westinghouse, Fundamentals of Nuclear Physics,. Chapter 7 N
I 1
4 I
i 2
, _ _ _ . . _ . . _ , , . _ - . _ _ . _ . _ _ - _ - - _ _ _ _ _ . . _ _ , _ - ,....m. . . - , _ . , _ .I
Figure 5.1 a
lb w
. i 3 '
iso Time
> tr$
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t=0 t:5 a -
4 l
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w 5
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t:0 t=5 Iime
i Figure 5.1 YN i.
T r !
l :
i=o 7., , its (Prempt Emp)0.25 j(Stable feriod)o.Z5 BC D0 '
- ~
(Stable,SttR = 0)0.25 (Prompt Dror) 0.2s v -.
i n
(Prompt beop)0.25 ye 81015 / (5%kker)o.25 u
3
<2 f 0+ompt ,Tump) 0.25 d
l t=0.
Tm ;
t=s i
5.03. ( 2. 0 ) -
i
'The . reactor ~ plant has been operating for six months at 100 %
power. Subsequently power is reduced to 50 %. Select one of the four choices below which best describes what happens.
l j (1) Stays the same (2) decreases by less than 50 %
(3) decreases by 50 %
(4) decreases by more than 50 %
I -(a) Approximatly how much is equilibrum xenon
- reactivity changed 7 (1.0) l (b) Approximatly how much is equilibrum samarium reactivity changed 7 (1.0) 30 (a) (2) (Decreased by approximatly 4-- %)
(b) (1) (No change, it stays the same.)
References:
Westinghouse, Reactor Core Control for Large. Pressurized Water Reactors, Chapter 4 Diablo Canyon, Operator Information Manual, Appendi:: R i
o 3-
. ---. ~ . , . _ . , _ . - . , . _ _ , , _ _- . --. , , . . , ._
o 5.04 (2.0)
The reactor has been operating at 100 % power for several weeks.
Which fission product poison, xenon or samarium, has a greater effect on reactivity ?
(a) Twelve hours after shutdown. (1.0)
(b) Thirty days after shutdown. (1.0)
(a) :enon (b) samarium
Reference:
Westinghouse, Reacter Core Control for Large Pressurized Water Reactorc, Chapter 4 4
,7__
- v i
5.05 (3.0) 3 :.
Following a loss of all off site AC power the plant is being cooled down by natural circulation using Emergency Procedure EP E-0.2 and E-0.3. Reactor coolant pressure is currently 1900 PSIG, the highest core exit thermocouple reads 590 F, hot. leg temperature indicates 545 F, and cold leg temperature indicates 510 F. -(Use the attached steam tables as appropriate. For calculations show your work and state your assumptions.)
(a) What is the RCS subcooling ? ( 1. 0 )
(b) At 'what pressure will voids begin to form in the ; '
reactor vessel ? (1.0) -
(c) What reaction would generate hydrogen gas in the RCS at high core temperatures ? ( 1. O V "'
s (a) 1900 PSIG corresponds to 630 F at saturation. 630-5hOF =
so E (MWDX\ MATLY corresponds ! +I7 PSIG to) 1533r (b) 590 F (hottect part of core)
(c) The circenium water reaction (evolves hydrogen at ,
high temperatures).
References:
Emergency Procedure EP E-0.3, Natural Circulation Cooldown with.
Steam Void in Vessel Westinghouse, Thermal-Hydraulic Principles and Applications to the Pressurized Water Reactor, Part~I (Chapter 2) and Part II, page 13-15 b
s
- )
s.
l i
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.g . g .
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w .\
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4 5.06 (3.0) -
- 1 l ,* \
e' Emergency di.usel generator 1-1 -io .being paralleled with bus H per e's Procedure OP J-6B to' load test the diesel.
t OperatirwlQ? Z-(a) Why i s' the diesel speed adjusted to obtain the synchroscope rotating slowly in the fast direction 9 just before shutting the diesel breaker ? (1.0)
\
(b) Hew'are the f ollowing electrical meter indications
' aHected if the output voltage controller of the y diysel generator i s' in'cr eased while it is paralleled with auxiliary power ?
~ N (0.5)
- 1. Generator voltace
- 2. Generator amps, (O.5)
- 3. Generctor KW ' t. (0.5)
- 4. Gener'hter
~.
VAR "lassume an initial 100 KVAR lead) (0.5) 3\
s .
.\ 0,
['f .y^ % 4 (a) Tc;drevent a reverse current trip and/or to cause the generator to pick up some loac when it is parallelcd.
y diesel a w 14 j 1 '
by , *t y, (b) 1. Voltage stays about the sarae.(OA. a)CRGASES SOME80 HAT)
- 2. Amps will increase.
- 3. KW Ni11 stay about the same.
s 4. VAR,wil1 3 - ~ - - -
dete eA se, ,
4 q
w y L..
References:
Operating Procedure OP J-6B:IV Diesel Generator 1-1 Manual
, ,O p e,r a t i e n s
(,
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g 4 z, %
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+
5.07 (2.0)
A small volume of air has been accidentally injected into the reactor . coolant system while it is at full power. State how the following quantities and parameters .be affected. (increase, decrease, stay the same) Briefly explain your answers.
(a) Hydrogen in the reactor coolant., (1.0)
[
(b) Reactor coolant Ph (1.0) l (a) Decrease. - ( 0. 5) Nitrogen and oxygen combine with hydrogen to form ammonia and water.
V '
(b) Increase ammonia F-rF4 (due .to f ormation of
_ { O F* % oAlfV M H V D$C WlD G (1. Q}
i
Reference:
Westinghouse, Radiation, -Chemistry and Corrocion Considerations for Nuclear Power P1 ant Application, Chapter 7 4
~
I i
e b- 9
! l p
L
__---_-_:-______ _ _ _ - - _ _ - _ _ _ _ - _ - . _ - _ _ _ . _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ f
- a s.
,1 s
ss 5.08 (3.0)
\
T(cold) is 545 F, core delta T is
/ While operating at 50 */. power, i 32 F, and RCS flow rate is 120 million Lbs/Hr.
,'y I , t
~~lFollowing a unit trip and loss of off-site power, a natural t'
circulation cooldown is establ,ished with T(cold) 505 F, core delta T 20 F. It is estimated'that core power is 7.68 x 10+7 Btu /Hr.
Assume Cp of water is 1.0 Btu /Lbm-F and use the attached steam tables,- if appropriate, to. answer the following questions. For calculations show your work'and state your assumptions.
} : >
(a) Before the unit trip, what is the initial core thermal' power insBtu/Hr ? (1.O)
(b) What is the natural circulation flow rate in Lbm/Hr ? (1.0)
(c) What is auxiliary.feedwater flow rate (Lbm/Hr) to the steam generators to maintain a constant T(ave')
if f eedwater at 70 F' (h = 40 Btu /Lbm) is used ? (1.0) 4 A LS 0 h cc epi 718tG
- . e (a) O=M: Cp :: delta T Q:/M ak Q= (120 :' 10+6) (1. 0) (32) n 3.84
--- - ---- x 10+9 Btu /Hr 0*'7b#lO [HV (c p _wg ,3 4 gyyjL, op}
e (b) Q = M :. Cp :. delta T 1
)
M = O/Cp x del ta T = 7. 68 :: 10+7/20 =
, ! T,
- 3. 8 4 :: 10+6 Lbm/Hr 4p e * ) . .
(c) O= M :: delta h theref ore M = O/ (h1 -- h 2 )
40 M= (7. 68 - :. 10+7) (1201 -M) = 7.68 x 10+7/1102 e (.G 15 0 I;l _- My, .. Lbs/Hr
Reference:
Wentinghouse', fhermal Hydraulic Principles and Applications to the Prescurized Water Reacter II,' Chapter 12
. . . O
[
\2 b' j L
t n ,
,h_ s' b M s
, FIGURE 5. 2.
- . !. 1 1 m : : = > = >
' 80 2 7
d[ d[ d( I '
)
i
-+; -
(< (<
a l y 2 3
- I %
c 3
=
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= . 8s
- 2 8#
dldlE= . W "s
= = sa 5 E e
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- (( ((
N/ 5/ %e M
T T
i 5.09 (3.0)
Refer to Figure 5.2. The plant is being heated up using four reactor coolant pumps for heat.
(a) What happens to the indicated amps on pump number 4 if the RCS temperature increases from 200 F to 400 F ? (1.0) m (b) What happens to reactor core delta pressure if pump number 4 is stopped ? (1.0)
(c) What happens to loop 2 flow rate if pump number 4 is stopped ? (1.0)
(a) Decreases (due to lower coolant density)
(b) Decreases (due to lower total flow rate)
(c) Increases (due to lower core delta pressure)
Referencec:
Westinghouse, Thermal Hydraulic Principles and Ap p l i c e ti crm +o the Prescurized Water Reactor II, Chapter 10 l
l
5.10 (2.0)
A heat balance by manual calculation has been completed in accordance with STP R-2B, Operator Feat Balance.
(a) If erosion has enlarged the feedwater venturis, (1.0) how is calculated feedwater flow affected ?
(b) If blowdown flow is omitted from the calculations, how is calculated power affected ? (1.0)
(a) Calculated flow is lower than actual (due to lower delte pressure).
(b) Cal cul ated power is higher than actual (due to less energv removed by blowdown, or higher energy removed by steam).
Reference:
Surveil l ance Test Procedure STP R-2D, Oper ator Heat Dalance 10
6.01 (3.5)
While firling and venting the RCS in accordance with OP A-2:I, the operator finds that he is unable to open letdown isolation valve (LCV 459) from the main control board.
(a) What are three of the four conditions that must be met in order to manually OPEN the letdown isolation valves from the main control board? (1.5)
(b) What condition must be met before manually CLOSEING the letdown isolation valves from the main control board? (1.0)
(c) What is the reason for the interlocks between the letdown isolation valves and the orifice isolation valves? (1.0)
(a) Threc out of four required. 0.5 pointe each.
- 1. All letdown crifice isolation valvet (G149 A, B and C) must be cloned. (0.7 7)
- 2. Pr escur i n er l evel 'must be > 17X.(0,7f)
-- - .- .- _ r u ._,,,-
-J 9 9 J _ 4 t 9
.A F,--- 3 (b) All letdcwn crifice icolation valven (8149 A, B and C) must be closed.
(c) To incur e the regenerative heat c:: changer always has Reactor Coolant System pressure in it (Also accept - To prevent steam flaching in the regenerative heat exchanger.)
Reference:
Diablo Canyon Lescon Plan B-la, Paragraph II.B.1.b(1) s 1
m
o 6.02 (1.5)
The feed pump speed control is designed to maintain a programmed pressure differential between the feedwater pump (s) discharge and driver steam header by controlling the speed of the turbine (steam governor speed changer).
What are the three reasons for this design?
- 1. Mai nt ai n feedwater control valves in a lincar range for better throttling characteristics. (0.5)
- 2. Reduce pump power requiremente at part load. (0.5)
- 3. Reduce the possibility of val ve plug crocion duc to cxcentive clonure at part Icad. (0,5)
Reference:
Dieblo Canyon Lescon Plan C-Pb, F*ar ag r aph II.D.1 l
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6.03 (1.5)
Steam generator program level is a function of turbine first stage pressure.
What are three significant reasons (bases) for programming steam generator level ?
- 1. Li mi t containment pressure en steam break incidt conte: n e c r< t .
(0.5)
- 2. Limit cooldcwn of the Reacter Coolant Syctem dering a cterm break. (0.5)
- 3. Miniminc the effectc of ch r i ni- and twell during ope r ati cn ol trancients. (0.5)
Also accept - Limit the neverity of a steam break at low power levels.
Reference:
Diablo Canyon Leccon Pl an C-Sb, Paragraph II.A.3.b.(3) 0 1
7 a
E
6.04 (2.0) are five factors or objectives that the Emergency Core There Cooling System is designed to attain.
Name four of the five objectives or factors.
Any 4 required. 0.5 each.
- 1. Cladding Temperature lecc than or equal to 2200 degrees F
- 2. Hydrogen generated s 1% of the maximum pocsibic
- 3. Total ox i dati on of clad < 1 7 7.
- 4. Maintain a gecmetry amenable to cooling
- 5. Long term core cooling capability
Reference:
Diabic Canyon Lesson Plan B-3, Paragraph I.E I
t 1
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a
6.05 (3.0)
There are several signals that cause different levels of containment isolation. Selecting only from High-high containment pressure, manual initiation, and Phase B isolation; which si gnal (s) would cause each of the following to occur in components associated with containment isolation ? (There may be more than one signal for each of the following isolations or closures.)
(a) Main steam line isolation valves and bypasses close. (1.0)
(b) Containment ventilation isolates. (1.0)
(c) Steam generator blowdown valves inside containment close. (1.0)
(r.o)
(a) High-High Containment Preccure ( ;_ . _ ;
s.--
r' L --r T. -,* r- - ; .~ e , in et (j.0)
(b) Manual Initiation (J. . '
w r= c . . rg (f.0)
(c) High-high Containment Preccuro ( : . _l
= _ . - . . , . , , . ,,
Reference:
Diablo Canyon Leccon Plan B-6c l
.i
6.06 (2.0)
The rod control system has rod stop and rod motion inhibit signals that affect in and out rod motion.
(a) What is the difference between the cause of a rod stop signal and the cause of a rod motion inhibit signal? (1.0)
(b) What type of rod motion, ie., in or out, does each signal prevent or not prevent? (1.0)
]
(a) Rod stop signal -- C au cced b ', plant conditionc ( 0. 5)
R o .d Moticn Inhibit cignal - Couced by f 11urc t. i th.. r e ."
centrol cyctem (0.5)
(b) Rod ctop cignal - Stapc r oc out motion, doet not s- + e p r c .'
i n--nc t i on (O. 5)
Red Nation Inhibit ciqnci - E. L oa c rod in and out nt1; (0.5'
Reference:
Diable Canyon Lect.on Plen A--S a , Paragraph 7.9.:. .
l i
6.07 (3.5)
Since No. 1-3 diesel generator can supply power to the 4KV Bus F of either Unit #1 or #2, it has_several features which are unique.
(a) If No. 1-3 diesel generator were in service on Unit #1's Bus F, but not as a result of a Safety Injection signal.on Unit #1, what would occur if a Safety Injection signal came in on Unit #2? (0.S)
(b) If No. 1-3 diesel generator were answering a Safety Injection signal on Unit #2 and a Safety Injection signal came in on Unit #1, what would be the initial response of the system? What must be done in order for the 1-3 diesel generator to shift to Unit #1? (1.5)
(c) In order.for No. 1-3 diesel generator to operate in the isochronous mode, how must the diesel engine MANUAL / AUTO selector switches in Units #1 and #2 be positioned? (0.5)
(d) How must the MANUAL / AUTO selector switches on
. Units #1 and #2 be positioned in order to block auto starts of No. 1-3 diesel generator? (0.5)
(e) If Unit #1 MANUAL / AUTO selector switch is in AUTO, and Unit #2 MANUAL / AUTO selector switch is in MANUAL and an auto-start signal comes in on Unit #
2, what will be the response of No. 1-3 diesel generator? (0.5)
(a) The diesel generator feeder breaker to Unit #1 would trip to allow trancfer to Unit #2. ( 0. 5)-
(b: -The No. 1-3 diesel generator would remain ccmmi tted to Unit M2 (0.75) until its (Unit M2) safety injection is recet (0.75), then it would chift to Unit #1.
(CM EsTHERuvuT)
(c) AUTO ,(0.5)
(d) MANUAL (0.5)
N oT' (e) No. 1-3 diesel generator will auto-start (0,5).
Reference:
Diablo Canyon Leccan Plan J-6b, Paragraph II.F,7.c, Diablo Ccnyon Unitt Mi and #2 differencer. book 7
n
6.08 (1.5) 4 Unit #1 and Unit #2 differ somewhat in core design and internaIs.
(a) The design of the reactor vessel internals of Unit
- 2 allows an increased mass of water in the plenum area. What are the two operational considerations
'that result from this increased mass of water? (1.0)
(b) Unit #1 has temperature monitoring instrumentation in the plenum area and Unit #2 does not. How does this additional input affect subcooling temperature indications during a natural circulation cooldown of Unit #1 if CRDM fans are not running? (0.5)
Unit #2 will be more restrictive on a (a) 1. Soak times on natural circulation cooldown. (0.5)
- 2. A larger inventory of water in the prorsurizer wou!d be required to collapse a void in the plenum of Unit #2. (0.5)
. eye _ Mp,[@,6 4 ,
gp gjgggggg yggg (o,q ("7EMfffeltBE (b) may indicate ,
'* 5) SUDthngs hiLE VORM ALLY MOT IU SEgysCG)
Reference:
Di abic Canyon Uni ts #1 and #2 differencoc book l
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6.09 (3.0) 1 multiple The Process Radiation Monitor System consists of plant channel s which monitor radiation levels in various operating systems.
a Select 6 out of the 8 monitors below and state the result of high radiation condition. State either alarm only, or describe the isolation that takes place. If an isolation takes place, valve numbers are not required, but give the general nomenclature for the valves that shut.
Any 6 required. 0.5 each.
(a) Containment Air Particulate Monitor (RE-11)
(b) Containment Radioactive Gas Monitor (RE-12)
(c) Residual Heat Removal Heat Exchanger Compartments Exhaust Duct Air Particulate Monitor (RE-13) 1 (d) Plant Vent Particulate Monitor and Backup (RE-28A and RE-28B)
)
(e) Condenser Air Ejector Gas Monitor (RE-15)
(f) Component Cooling Liquid Monitor (RE-17A and 178)
(g) Steam Generator Blowdown Sample (RE-19)
(h) Plant Vent (Iodine Monitor) (RE-24)
Any 6 required. 0.5 each
, (a) Containment ventilation isolation (b) Containment ventilation isolation (c) Alarm only (d) Containment ventilation isolation (e) Alarm only i
(f) Surge tank vent (RCV-16) closes (g) Isolate blowdown -
(h)
Reference:
Diablo Canyon Lescon Plan G-4, Paragraph II.D 9
6.10 (1.5)
The 4 KV system provides electrical power to motors less than 3000 hp and to the 480 V load center transformers.
(a) What is the normal source of power to the 4 KV busses ? (unit operating) (0.5)
(b) What is the alternate source of power to the 4 KV busses ? (unit chutdown) (0.5)
(c) What is the emergency source of power to the 4 KV busses ? (0.5)
(a) Au:<ili ary t r ans f ormcr (0."
(0',
(b) Otcrtup trancfermer (c) Di etel generators ( 0. G )
Reference:
Diablo Canycn Lencan Plan J -62, Par agr ap:t i I I . r: .1. c .
l 10
6.11 (2.0)
The reactor trip bypass breakers are physically located alongside their respective reactor trip breaker.
(a) What is the purpose of the reactor trip bypass (1.0) breakers?
(b) How would the reactor trip breakers and reactor trip bypass breakers respond if bypass A and simultaneously? (1.0) bypass B were racked in the solid state protecticn ryntec (a) To allow t est i rig of without tripping the reactor.
(b) The reactor trip breal'ert would trip.
The reactor trip bypass brealers would trip.
Ftef eren ce: Di abl o Ganyon Letron Plan e-3a, r'ar ag r ap h 6. fi 11
7.01 (2.5)
Emergency boration can be required for a number of reasons. In each of the situations listed below, at what point may emergency boration be stopped?
(0.5)
(a) One stuck rod.
(0.5)
(b) Violation of rod insertion limits.
achieved inadvertently below the rod (c) Criticality (0.5) insertion limits.
margin less than that required by (d) Shutdown (0.5)
Technical Specifications.
(e) Uncontrolled cooldown following a reactor trip. (0.5)
(3)
M.cWR. REMiRED[0%~ENCV NRATico uoi- (26GMED 00. Tit.
r 2 (2cci 39A9VC K) -- 6 .n^ . (0.5)
(0,5)
(b) borate until rods are obove red insertirn limitc.
reactor ccolant system baron (c) Dorate to increane concentration by 100 ppm. (0,5)
Dorate until sufficient shutdown margin achicvea. (0.5)
(d)
(c) Derate an necessary to regain control. (0.5)
Reference:
Diablo Canyon Power Plant Unit No ( rs ) i cod 2 Operating Prccedure OF AP-6 1
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7.02 (2.5)
Hagan controllers are used to send demand signals to valves
+ and position or function.
other equipment which regulate their According to OP O-2, there are two types of power to these controllers, automatic and manual power.
(a) If a Hagan controller i s operating in AUTO and
. automatic power is lost, how will the controller l (0.5) i respond?
a Hagan controller i s operating in AUTO and (b) If manual power is lost, what is the immediate i (0.5) response of the controller?
(c) If a Hagan controller is operating in AUTO and manual power is lost, how will the controller respond once manual power is restored? (0.5) 4
- (d) -If a Hagan controller is operating in MANUAL and I manual power is lost, how will the controller (0.5) j respond?
1 (e) a Hagan controller is operating in MANUAL and j' If the controller manual power is lost, how will I respond once manual power is restored? (0.5) 1 i
(a) The centroller will go to MANUAL (and stay there.) (O.S) i (b) The controller will go to AUTO-HOLD. (0.5) i '
+ (c) The controller will go to MANUAL. (0.5) i The controller will go to AUTO- HOLD. (0.5)
(d)
(0,5)
J (e) The contro11cr will trannfer back to MANUAL.
4 4
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Reference:
' Diablo Canyon Power Plc.ht No(c) 1 and 2 Operating Procedure GP j 0-2 i ,
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7.03 (4.0)
According to OP O-3, there are numerous items that require prompt notification of the SHIFT CHEM / RAD TECH.
(a) Excluding the notification required of RMS channels .that have been determined to be inoperable, name four events for which the SHIFT (2.0)
CHEM / RAD TECH must be notified.
(b) In what two logs may this notification be i (1.0) documented?
(c)' What are two important details that the documentation should contain? (1.0)
(a) Any f our required. 0.5 each.
- 1. Reactor startup in progress
- 2. Reactor shutdown in progrens.
- 3. Reactor power change greater that 15% in one hcur.
- 4. Refueling canal in being flooded.
- 3. 51 accumulator volume increase of greater than 12% on vertical board 1 (alco accept - SI accumulator volume increase of greater than 1% of an accumulator'n total tank volume.)
(b) In the Shift Foreman (0.5) or Control Operator *n (0,5) log.
(c) Any two required. 0.5 each.
- 1. The nature of the notification.
- 2. The name af the SHIFT CHEM / RAD TECH notified.
- 3. The time of the notification.
Reference:
Diablo Canyon Power Plant Unit No(s) 1 and 2 Operating Procedure OP O-3 3
9
+
7.04 (2.0)
OP L-5 states that the reactor coolant pump #1 seal bypass valve (8142) should NOT be opened unless either the pump bearing temperature or the #1 seal leakoff temperature approaches its alarm level.
What are the four additional conditions that must be met in order (2.0) to open valve 8142?
- 1. 1000 prig > RCE prescure -' 100 pci g. (0.5)
- 2. #1 seal leakoff valves are open. (0,5)
- 3. til r.e a l leakoff flowrate is x 1 gpm. (0.5)
- 4. Seal injection water ficwrate to each pump ic 6 gpm. (0.5)
Reference:
Diablo Canyon Poveer Plant Unit No(s) 1 and 2 Operating Pr ocedure GP L-5 J
/1
7.05 (2.0)
The operator selects the two nuclear instrument channels that he For the following conditions wants recorded by recorder NR-45. (ie.,
reactor startup, what type (s) of channel (s) during a Source, Intermediate or Power) and how many, should the operator select to be recorded by NR-45?
(1.0)
(a) Prior to withdrawal of the control banks.
(1.0)
(b) Reactor power stable at 10 -8 amps.
range (0.5) tind one i n t er nied i at e range (0.5?
(a) One cource channel.
range (0.5) and one powcr - an g r. 0.54 (b) One i nt er n.edi ct e channel.
Retcrence:
Dinblo Canyon Power P1cnt Unit No(c) 1 and 2 Operat;ng Pr o:_- jor e OP L-2
c 7.06 (2.0)
OP AP-8 requires three immediate actions if the control room must be evacuated. One of these is to proceed to the Hot Shutdown Panel.
What are the other two, and how is each verified?
- 1. Manuall y trip the reactor before leaving the control room.
Verify by all contrcl rod bottom lights on and reactor power en NIS decreasing.
- 2. Verify tripped or manually trip the turbine before le.,'ing the control room.
Verify al1 four ctop valven closed on the turbine EH panel.
Reference:
Diablo Canyon Power Pl ant Unit No (s) 1 and 2 Operating Pr ctedur e OP AP-G ,
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7.07 (3.0)
There are certain minimum manning levels-required in the control room during reactor startups.
(a) What are the minimum manning requirements (licensed operators only) in the . control room a routine reactor startup.? (1.5)
-during (b) Prior to beginning a reactor startup following a reactor trip, whose approval must be obtained? (0.5)
(c) Prior to beginning a startup following a reactor trip, one of several managers must be in the (1.0) control room. Name two of them by position.
(a) One Reactor Operator (0. 5) , ono Senior Reactor Oper star and one additional licensed cperater (0.5) (RO or (0.5),
SRO).
(b) Plant Superintendent (0.5} (or his delegate)
ONE -" '
(1,0)
(c) Any - required.
- 1. Techni c al Managcr
- 2. Engineering Manager
- 3. Senior Power Production Engineer
- 4. Power Producticn Engineer (&ft. (AE60est, fMN4F6th
Reference:
Diablo Canyon Power Pl ant Unit No(s) 1 and 2 Operating Proceduro OP L-2
-7
7.08 (2.5)
EP' E-3 describes indications of a steam generator with ruptured tubes or a fault (other than ruptured tubes).
-(a) Name three process radiation monitors that may be checked in order to detect a steam generator with a tuhg tuntute. (1.5)
(b) -If checking a steam generator for a fault (other than tube rupture), what trend or condition is looked for regarding steam generator pressure? (1.0)
(a) 1. Steamline radiation. (0.5)
- 2. Steam jet air ejector radiation. (0.5)
- 3. Steam generarcr blowdown radiation. (0.5)
(b) 1. Steam generator pressure decreating (in an uncontrolled manner.) (0.5)
- 2. Steam generator completely depressurized. (0.5)
Reference:
Diablo Canyon Power Plant Unit No(s) 1 and 2 Emergency Procedure EP E-O O
l 7.09 (1.0) to EP E-1.2, reinitiation of safety injection would According require manual starting of the ECCS pumps.
What are the two safety injection reinitiation criteria? Assume (1.0) normal containment conditions.
R r c, cubccoling Iece than 20 degrees F. (0.5) 1.
levc1 cannot be maintained above 4%. (0.5)
- 2. Pressurizar
Reference:
Procedure Diabic Canyc.n Powcr Plant Unit No(s) 1 and 2 Emergency EP E-1.2 1
9
7.10 (3.5) the following questions in accordance with the Respond to requirements of EP FR-S.1 - Nuclear Power Generation /ATWS.
(a) What two observations / indications are necessary to (1.0) verify a reactor trip?
(b) If the reactor does not trip manually, what is the next step in attempting to trip the reactor? (0.5)
(c) What four steps are necessary to emergency borate (2.0) through emergency borate valve 8104?
(a) 1. Reactor trip and bypinc breskert open. (0.5)
- 2. Rod bottom lights lit. (0.G)
(b) Manually deenergine the 4 R 0ti tr u s s er- f eedi n g t ht- r or.1 dr v: PL n?tc. (O.D)
(c) 1. Etart charging pumpt. (0.5)
- 2. Operate the bcric acid t'~an uf er pucp at Icw spct J . (0,5)
- 3. Open valve 8104 (emt-rgency borate val vc) . /O.5)
- 4. Manually increasc charging to ?S gpm. (0.5)
Referenec:
Diablo Canyen Pcuer Plant Unit Nc(c) 1 and 2 En.ergency Pr ac t. c . t-EP FP-S.1 10
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1 8.01 (3.0) ,
i i The CODE OF FEDERAL REGULATIONS (10 CFR 55) defines the general provisions for Operator's Licenses. In accordance with these regulations:
1 (a) What are the " controls" of a nuclear facility? (1.0) , ,
-,,. 4-is an individual deemed to be operating the (1.0) i 1 (b) When i controls of a nuclear facility?
(c) Who may operate the " controls" of a nuclear facility without an Operator License? (J}4 0) +.
! u
.A
. (a) " Controls" - apparatus and mechanisms the manipulation of which directly affects the reactivity or power level of the r reactor.
,! ' (b). An indi vi dual i s deemed to operate the controls of
~
a j facility' if he (1) directly manipulates the controls or (2) i directs another to manipulate the controls.
3 j (c) An- individual may manipulate the controls as a part of his training to qualify for an operator license under the . y, direction and i n the prononce of a licensed operator or conier operator.
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I 10 CFR 55 l
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- y-8.02 (3.Of 4
In accordance with the Technical Sp'?cification f or Administrative s
V controls, Section 6.2.2 " Plant Staff":
a
, y-(* g ,ga) What is the minimum crew composition required with (1.0) 4', both units in Mode 3 ?
~;, . ,+ ,
4,
, h (b) What is the minimum crew composition required with (1.0)
\' both units in Mode 5 ?
\ (d) What two additional manning requirements, other (1.0) than the Shift Crew complement, must be met any time there is fuel (!n the reactor?
g ,
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(a) (O.20 each)
$ (
1 - Eh2ft Super vi cor ' (' (Licensed SRO) f 6' . ,m \
Ns
1 - Eenior Pgacter Operator (Licensed SRO) s ( 3 - Reactor Operatorc (Licenced RO) g 3 - Auxiliary Operators '
m' '
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1 - Shi f t, Tec,brb cal Advisor
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(b) (0.33 ec.ch) s
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('M (0.50dgd t-s , s 1
1'3 Heal th Phvsi cs Tercnici an 5 - Site Fire Erigade
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Tech Spec 6.2.2 M- g 1
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Each Accumulator is required by Technical Specification 3/4.5.1 to be demonstrated operable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying borated water volume, nitrogen cover-pressure, and accumulator isolation valve position. Some extensions of the basic interval are allowed by the Technical Specification.
Records show that this was done on:
November 11 at 0000 November 11 at 1500 November 12 at 0000 November 12 at 1400 November 13 at 0500 (a) 'What is the maximum allowable interval between (1.0) accumulator system Surveillances?
(b) When is the next surveillance due? (1.0)
(a) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> + 25%-(3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) = 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> Not to e::ceed 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> + 25 % (3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) = 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> for three survoillance intervals.
(b) November 13 at 1500 0500 + 15 = 2000 but cannot exceed 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> in three surveillances thus 11/13 at 1500.
Tech Spec 3/4.5.1 & 4.0.2 t
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8.04 (1.0) the Technical Specifications at what RCS As defined in temperature do you change from Mode 3 to mode 4 ?
350 degrees F 7
Tech Spec definitions
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8.05 (2.O)
A plant safety limit has been violated while in Mode 1. Technical Specifications require that two actions shall be accomplished within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. What are these two actions?
(a) The uriit shalI be p1 aces in HOl STANDBY (b) The NRC OPERATIONS CENTER shall be notified by telephonc.
Tech Spec 2.1 & 6.7 l{
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L 8.06 (3.0)
.i Shutdown Margin is defined in the Technical Specifications as the amount of reactivity by which the reactor is instantaneous subtritical.
(a) What is the required SHUTDOWN MARGIN when in COLD SHUTDOWN ?
(b) What is the required SHUTDOWN MARGIN when above 200 degrees F 7 ,
(c) Why is there a difference in part a and b above 7 4
I-(a) 1.0% delta K/K (b) 1.6% delta K/K (c) Because the effect of the moderator temperature coefficient is very small at the lower temperatures reducing the reacti vi ty ef f ects of a rapid cooldown.
Tech Spec 3/4.1.1, Bases 3/4.1.1 i
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8.07 (2.0)
Using the classifications defined in Proceedure EP G-1, Accident Classification and Emergency Plan Activation, as given below, classify the described situations.
Classifications:
Unusual Event Alert Site Emergency General Emergency Situations:
(a) Total loss of heat sink. (0.5)
(b) Inadvertent SI actuation. (0.5)
(c) S/G tube rupture of 750 gpm. (0.5)
(d) Release of gas decay tank that exceeds Technical (0.5)
Specification limits by 10%.
(a) General Emergency (b) Unusual Event (c) Alert (d) Unusual Event Procedure EP G-1 7
8.08 (2.0)
In accordance with federal regulation (10 CFR 20);
(a) What is a RADIATION AREA ? (1.0)
(B) What is a HIGH RADIATION AREA ? (1.0)
(a) Arca (acccccible to personnel) where major part of the bcci could receive:
5 mrem in one hour or 100 mrem in 5 days (b) Area (accessible to personnel) where major part of the body could receive:
100 mRen in one hour 10 CFR 20 8
l.
8.09 (2.0) i Indicate if each of the following is TRUE or FALSE.
(a) The plant is in mode 3. A condensate storage tank level of 195,000 gallons meets Technical Specification requirements. (0.5) 4 i
i (b) The shutdown banks are allowed to be included in
! shutdown margin calculation even though they are I withdrawn from the core. (0.5)
(c) The pl' ant is in mode 1. The raw water resevoir usable volume of 200,000 gallons is within Technical Specification limits. (0.5)
(d) There is no Technical Specification addressing the gaseous radwaste system. (0.5)
(a) True (178,000 gallons required)
, (b) True (see. definition)
(c) False (270,000 gall ons required)
(d) False - (ma::imum of 10+5' curies / tank, oxygen alarm and trip) .
I 1
i
~
i l Ref erence:
Technical Specifications, 3.7.1.3, definition 1.31, 3.7.9.1, 3.11.2.5, and 3.11.2.6 4
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8.10 (3.0)
There are requirements in the Technical Specifications for both off-site and on-site electrical power sources.
(a) What major A.C. electrical power sources are required by Technical Specifications in order to enter mode 4 ? (2.0)
(b) If the motor operated disconnect was inoperable, would this affect the operability requirements stated in part (a) above ? (1.0)
(a) (1) Two independtnt circuits (one with del ayed eccent) between the off-site transmission netwoprk and the on-cite clast IE distribution system, and (2) Three separate and independent diesel generators.
(b) It could. If the MOD was inoperable and in the c l tc ed pccition we would not have the 500KV switchyard av r.i l ab i c az one of the "cff-site transmission network" cources.
Reference:
Technica) Speci f 2 cat i on 3/4.6.1 i
10
.. e 8.11 (2.0)
During refuling the Technical Specifications prohibit loads in excess of 2500 pounds from being moved ofer fuel assemblies in the spent fuel pool. What are two reasons for this limit ac stated in the bases for the Technical Specifications ?
(c) Li mi t the activity releace in the event of load drop to that of a single fuel element. (1.0) and (b) Limit possible fuel distortion in the fuel storage racks 50 as not to result in a critical aray. (1,0)
Reference:
Technical Specification, 3/4.9.7, Crane Travel e
11